Method of treating waste from nuclear fuel handling facility and apparatus for carrying out the same

ABSTRACT

A waste treatment apparatus treats radioactive contaminated waste from a nuclear fuel material handling facility to decontaminate the radioactive contaminated waste by using an electrolytic molten salt, and reuses the electrolytic molten salt so that any effluents are not produced. Radioactive contaminated waste ( 10 ) from a nuclear fuel material handling facility is subjected to electrolysis by a molten salt electrolysis unit ( 20 ) to decontaminate the waste ( 10 ). The used salt ( 16 ) used for decontaminating the waste ( 10 ) is filtered to separate nuclear fuel materials ( 19 ) from the used salt ( 16 ). The filtered salt ( 18 ) is reused by the molten salt electrolysis unit ( 20 ). The salt adhering to the decontaminated waste ( 12 ) is recovered by an evaporating unit ( 59 ), and the recovered salt ( 15 ) is reused by the molten salt electrolysis unit ( 20 ).

RELATED APPLICATIONS

This application is a divisional of U.S. patent application Ser. No.09/393,317, filed on Sep. 10, 1999 now U.S. Pat. No. 6,299,748.

BACKGROUND OF THE INVENTION

1. Field of the Invention

The present invention relates to a method of treating electricallyconductive waste contaminated with nuclear fuel materials and disposedfrom a nuclear fuel handling facility, and an apparatus for carrying outthe method. More specifically, the present invention relates to a methodof treating contaminated metallic waste produced when a nuclear fuelhandling facility is dismantled, such as waste steel materialscontaminated with nuclear fuel materials, or an adsorbent used foradsorbing nuclear fuel materials mounted in a nuclear fuel handlingfacility, and an apparatus for carrying out the method.

2. Description of the Related Art

FIG. 20 is a typical view of an apparatus for carrying out anelectrolytic polishing process generally used for decontaminating wastecontaminated with radioactive substances, such as nuclear fuelmaterials, (hereinafter referred to as “radioactive contaminated waste”)by electrolysis. As shown in FIG. 20, a radioactive contaminated waste 3is held by a holding device 2 and is immersed in an electrolyticsolution contained in an electrolytic vessel 1 of a stainless steel. Theradioactive contaminated waste 3 functions as an anode. A cathode 4 isimmersed in the electrolytic water solution 5. When the radioactivecontaminated waste 3 is a stainless steel waste, a phosphoric acidsolution is used as the electrolytic water solution 5 serving as a bath.When the radioactive contaminated waste 3 is a carbon steel material, asulfuric acid solution is used as the electrolytic water solution 5. Theholding device 2 and the cathode 4 are connected to a dc power supply 6.When a dc voltage is applied across the holding device 2 and the cathode4 by the dc power supply 6, the radioactive contaminated waste 3functions as an anode. A surface layer of the radioactive contaminatedwaste 3 dissolves in the electrolytic water solution 5 simultaneouslywith coming off of radioactive contaminants adhering to the radioactivecontaminated waste 3. Part of substances came off the radioactivecontaminated waste 3 remains in the electrolytic water solution 5 andthe rest is precipitated in sludge 8 on the bottom of the electrolyticvessel 1. Hydrogen 7 is produced on the cathode 4 of a stainless steel.

Generally, when decontaminating a radioactive contaminated waste by theelectrolytic polishing process using the electrolytic water solution 5as a bath, current is unable to flow uniformly over the entire surfaceof the radioactive contaminated waste when the radioactive contaminatedwaste has a complicated shape because the resistance of the bath ishigh. Consequently, the decontaminating effect of the electrolyticpolishing process is reduced for some portions of the radioactivecontaminated waste. If a high current is supplied to the bath to enhanceelectrolytic processing speed, heat is generated in the bath due to thehigh resistance of the bath. Hydrogen 7 produced on the cathode 4 duringthe electrolytic polishing process cause problems in safety. It isdifficult to remove radioactive substances accumulated in the wasteelectrolytic water solution 5, particularly, radioactive substancesdissolved in the waste electrolytic water solution 5 from the wasteelectrolytic water solution 5. The waste electrolytic solution 5 cannotbe reused and becomes an additional radioactive contaminated waste.Thus, the total amount of radioactive contaminated waste increases.

The present invention has been made to solve those problems and it istherefore an object of the present invention to provide a method oftreating waste from a nuclear fuel handling facility, capable of easilydecontaminating a contaminated waste having a complicated shape, notdischarging any effluent, capable of repeatedly using an electrolyticsolution and not producing additional waste.

Another object of the present invention is to provide an apparatus forcarrying out the foregoing method.

SUMMARY OF THE INVENTION

According to a first aspect of the present invention, a method oftreating electrically conductive waste contaminated with nuclear fuelmaterials from a nuclear fuel handling facility comprises a molten saltelectrolysis process for removing the nuclear fuel materials adhering toa surface of the waste by immersing the waste in a molten salt todissolve a surface layer of the waste electrochemically in the moltensalt; and a filtering process for filtering the molten salt used in themolten salt electrolysis process to extract the nuclear fuel materialsremoved from the surface of the waste and accumulated in the molten saltfrom the molten salt. The molten salt filtered in the filtering processis reused in the molten salt electrolysis process.

Preferably, the method further comprises an evaporation process forremoving the molten salt adhering to a surface of the waste processed bythe molten salt electrolysis process and taken out of the molten salt byheating the waste so that the molten salt adhering thereto evaporates.The molten salt recovered in the evaporation process is reused in themolten salt electrolysis process.

Preferably, the method further comprises a cleaning process for removingthe molten salt adhering to the waste processed by the molten saltelectrolysis process and taken out of the molten salt by a cleaningliquid, and an evaporative drying process for drying the molten saltcontained in the cleaning liquid by evaporating the cleaning liquid usedin the cleaning process. The molten salt recovered in the evaporativedrying process is reused in the molten salt electrolysis process, andthe cleaning liquid evaporated in the evaporative drying process isreused in the cleaning process.

Preferably, in the molten salt electrolysis process, the molten salt andthe waste immersed in the molten salt are moved relative to each otherto remove the nuclear fuel materials from the surface of the waste.

Preferably, in the molten salt electrolysis process, the waste iscontained in a basket serving as an electrode for an electrolysis andthe basket is vibrated in the molten salt.

Preferably, in the molten salt electrolysis process, the waste iscontained in a basket serving as an electrode for an electrolysis andthe basket is rotated in the molten salt.

Preferably, in the molten salt electrolysis process, the molten metal isspouted against the waste immersed in the molten salt.

Preferably, a liquid metal, which is in a liquid phase at a temperaturehigh enough to maintain the molten salt in a molten state, is placed inthe molten salt as an electrode for the molten salt electrolysisprocess.

Preferably, when the nuclear fuel materials are oxides, the methodfurther comprises a reducing process for reducing the nuclear fuelmaterials to metals before subjecting the waste to the molten saltelectrolysis process.

Preferably, in the reducing process, the nuclear fuel materials arereduced to metals by making the nuclear fuel materials react with areducing agent.

Preferably, the reducing process comprises immersing the wastecontaminated with the nuclear fuel materials in a reducing molten salt,supplying a reducing agent into the reducing molten salt, applying avoltage that will not cause a decomposition of the reducing molten saltacross an anode and a cathode immersed in the reducing molten salt toregenerate the reducing agent reacted with the nuclear fuel materials.

Preferably, the reducing process comprises immersing the wastecontaminated with the nuclear fuel oxides in a reducing molten salt,reducing the nuclear fuel oxides to metals by applying a voltage acrossan anode and a cathode immersed in the reducing molten salt for anelectrolytic reduction.

According to a second aspect of the present invention, a method oftreating an electrically conductive waste contaminated with nuclear fuelmaterials from a nuclear fuel handling facility comprises a reducingprocess for reducing the nuclear fuel materials to metals; a thermalmelting process for producing a molten salt by heating and melting themetals produced by reducing the nuclear fuel materials and the waste;and a molten salt electrolysis process for recovering the metalsproduced by reducing the nuclear fuel materials and contained in themolten salt by applying a voltage across an anode and a cathode immersedin the molten salt so that the metals produced by reducing the nuclearfuel materials are deposited on the cathode.

Preferably, a chloride or a hydride having a same kind of cation as thatof the molten salt is added to the molten salt to lower the meltingpoint of the molten salt so that an operating temperature of the moltensalt in the molten salt electrolysis process is lowered.

Preferably, the method further comprises a cleaning process forseparating the nuclear fuel materials from the waste by cleaning thenuclear fuel materials deposited on the cathode in the molten saltelectrolysis process and the waste with a cleaning liquid to dissolvethe waste in the cleaning liquid; and an oxidation process forconverting the nuclear fuel materials separated from the waste by thecleaning process into oxides by oxidizing the nuclear fuel materials;wherein the waste is an adsorbent used in the nuclear fuel materialhandling facility.

Preferably, the method further comprises an evaporative drying processfor drying the adsorbent contained in the cleaning liquid by evaporatingthe cleaning liquid used in the cleaning process. The cleaning liquidevaporated by the evaporative drying process is reused in the cleaningprocess.

According to a third aspect of the present invention, an apparatus fortreating an electrically conductive waste contaminated with nuclear fuelmaterials from a nuclear fuel handling facility comprises a molten saltelectrolysis unit for removing the nuclear fuel materials adhering to asurface of the waste by immersing the waste in a molten salt to dissolvea surface layer of the waste electrochemically in the molten salt; afiltering unit for filtering the molten salt used by the molten saltelectrolysis unit to extract the nuclear fuel materials removed from thesurface of the waste and accumulated in the molten salt from the moltensalt, and a molten salt return line for returning the molten saltfiltered by the filtering unit to the molten salt electrolysis unit.

Preferably, the apparatus further comprises an evaporation unit forremoving the molten salt adhering to a surface of the waste processed bythe molten salt electrolysis unit and taken out of the molten salt byheating the waste so that the molten salt adhering thereto evaporates,and a molten salt return line for returning the molten salt removed fromthe surface of the waste by the evaporation unit to the molten saltelectrolysis unit.

Preferably, the apparatus further comprises a cleaning unit for removingthe molten salt adhering to the waste processed by the molten saltelectrolysis unit and taken out of the molten salt by a cleaning liquid,and an evaporative drying unit for drying the molten salt contained inthe cleaning liquid by evaporating the cleaning liquid used by thecleaning unit, a molten salt return line for returning the molten saltrecovered by the evaporative drying unit to the molten salt electrolysisunit, and a cleaning liquid return line for returning the cleaningliquid evaporated by the evaporative drying unit to the cleaning unit.

Preferably, the molten salt electrolysis unit is provided with a drivingmechanism for moving the molten salt and the waste immersed in themolten salt relative to each other.

Preferably, the molten salt electrolysis unit is provided further with abasket capable of containing the waste and serving as an electrode foran electrolysis, and the driving mechanism vibrates the basket in themolten salt.

Preferably, the molten salt electrolysis unit is provided further with abasket capable of containing the waste and serving as an electrode foran electrolysis, and the driving mechanism rotates the basket in themolten salt.

Preferably, driving mechanism includes a spouting means for spouting themolten salt against the waste immersed in the molten salt.

Preferably, the molten salt electrolysis unit is provided with anelectrode formed from a liquid metal which is immersed in the moltensalt and is in a liquid phase at a temperature high enough to maintainthe molten salt in a molten state.

Preferably, when the nuclear fuel materials are oxides, the apparatusfurther comprises a reducing unit for reducing the nuclear fuelmaterials to metals.

According to a fourth aspect of the present invention, an apparatus fortreating an electrically conductive waste contaminated with nuclear fuelmaterials from a nuclear fuel handling facility comprises a reducingunit for reducing the nuclear fuel materials to metals, a thermalmelting unit for producing a molten salt by heating and melting themetals produced by reducing the nuclear fuel materials and the waste,and a molten salt electrolysis unit for recovering the metals producedby reducing the nuclear fuel materials and contained in the molten saltby applying a voltage across an anode and a cathode immersed in themolten salt so that the metals produced by reducing the nuclear fuelmaterials are deposited on the cathode.

Preferably, the waste is an adsorbent used for adsorbing the nuclearfuel materials in the nuclear fuel handling facility, and the apparatusfurther comprises a cleaning unit for separating the nuclear fuelmaterials from the waste by cleaning the nuclear fuel materialsdeposited on the cathode of the molten salt electrolysis unit and thewaste with a cleaning liquid to dissolve the waste in the cleaningliquid, and an oxidation unit for converting the nuclear fuel materialsseparated from the waste by the cleaning unit into oxides by oxidizingthe nuclear fuel materials.

Preferably, the apparatus further comprises an evaporative drying unitfor drying the adsorbent contained in the cleaning liquid by evaporatingthe cleaning liquid used by the cleaning unit, and a cleaning liquidreturn line for returning the cleaning liquid recovered by theevaporative drying unit to the cleaning unit.

BRIEF DESCRIPTION OF THE DRAWINGS

The above and other objects, features and advantages of the presentinvention will become more apparent from the following description takenin connection with the accompanying drawings, in which:

FIG. 1 is a block diagram of a waste treatment apparatus in a firstembodiment according to the present invention for treating waste from anuclear fuel handling facility;

FIG. 2 is a schematic longitudinal sectional view of a molten saltelectrolysis unit included in the waste treatment apparatus shown inFIG. 1;

FIG. 3 is a schematic longitudinal sectional view of an evaporating unitincluded in the waste treatment apparatus shown in FIG. 1;

FIG. 4 is a flow chart of a waste treatment method to be carried out bythe waste treatment apparatus shown in FIG. 1;

FIG. 5 is a schematic longitudinal sectional view of a molten saltelectrolysis unit in a first modification of the molten saltelectrolysis unit included in the waste treatment apparatus shown inFIG. 1;

FIG. 6 is a schematic longitudinal sectional view of a molten saltelectrolysis unit in a second modification of the molten saltelectrolysis unit included in the waste treatment apparatus shown inFIG. 1;

FIG. 7 is a schematic longitudinal sectional view of a molten saltelectrolysis unit in a third modification of the molten saltelectrolysis unit included in the waste treatment apparatus shown inFIG. 1;

FIG. 8 is a block diagram of a waste treatment apparatus in a secondembodiment according to the present invention for treating waste from anuclear fuel handling facility;

FIG. 9 is a schematic longitudinal sectional view of a cleaning unitincluded in the waste treatment apparatus shown in FIG. 8;

FIG. 10 is a flow chart of a waste treatment method to be carried out bythe waste treatment apparatus in the second embodiment;

FIG. 11 is a block diagram of a waste treatment apparatus in a thirdembodiment according to the present invention for treating waste from anuclear fuel handling facility;

FIG. 12 is a schematic longitudinal sectional view of a reducing unitincluded in the waste treatment apparatus shown in FIG. 11;

FIG. 13 is a flow chart of a waste treatment method to be carried out bythe waste treatment apparatus in the third embodiment;

FIG. 14 is a schematic longitudinal sectional view of a reducing unit ina first modification of the reducing unit included in the wastetreatment apparatus shown in FIG. 11;

FIG. 15 is a schematic longitudinal sectional view of a reducing unit ina second modification of the reducing unit included in the wastetreatment apparatus shown in FIG. 11;

FIG. 16 is a block diagram of a waste treatment apparatus in a fourthembodiment according to the present invention for treating waste from anuclear fuel handling facility;

FIG. 17 is a schematic longitudinal sectional view of a molten saltelectrolysis unit included in the waste treatment apparatus shown inFIG. 16;

FIG. 18 is a flow chart of a waste treatment method to be carried out bythe waste treatment apparatus in the fourth embodiment;

FIG. 19 is a schematic longitudinal sectional view of a reducing unit ina modification of the reducing unit included in the waste treatmentapparatus shown in FIG. 16; and

FIG. 20 is a schematic longitudinal sectional view of an electrolysisvessel of assistance in explaining a conventional electrolytic polishingfor decontaminating a contaminated waste.

DESCRIPTION OF THE PREFERRED EMBODIMENTS

A waste treatment apparatus in a first embodiment according to thepresent invention and a waste treatment method to be carried out by thesame waste treatment apparatus will be described hereinafter.

Nuclear fuel handling facilities include uranium mining facilities,uranium refining facilities, conversion plants, enrichment plants,nuclear fuel processing plants, nuclear reactors, reprocessing plants,waste disposal facilities, and transportation facilities fortransporting nuclear fuel materials between those facilities and plants.

Waste from nuclear fuel handling facilities includes various steelmaterials that are produced when nuclear fuel handling facilities aredismantled, and adsorbents which are used for arresting nuclear fuelmaterials in nuclear fuel handling facilities. The waste treatmentapparatus in the first embodiment is suitable for treating contaminatedmetal wastes, such as contaminated steel materials, or contaminatedmetal waste cut into small pieces by a pretreatment process.

Nuclear fuel materials include uranium, uranium ores, uranium oxides,uranium chloride, uranium fluoride, uranium hydride, uranium nitrate anduranium sulfate.

Referring to FIGS. 1 and 2, the waste treatment apparatus has a moltensalt electrolysis unit 20 for decontaminating radioactive contaminatedwaste 10 from a nuclear fuel handling facility by molten saltelectrolysis using a molten salt 24. The salt 24 adheres to thedecontaminated waste 12 decontaminated by the molten salt electrolysisunit 20. The salt 24 adhering to the decontaminated waste 12 isseparated from the waste 12 by an evaporation unit 59. The evaporationunit 59 melts and evaporates the salt 24 by heating the salt 24 adheringto the waste 12 at a temperature not lower than its melting point toseparate the salt 24 from the waste 12. The evaporation unit 59 is aknown evaporation device used in chemical engineering. The wastetreatment apparatus has a recovered salt return line 53 for returningthe recovered salt 15 separated from the decontaminated waste 12 andrecovered by the evaporating unit 59 to the molten salt electrolysisunit 20. The recovered salt return line 53 may be of either a transferpipe type or a conveyor type.

The waste treatment apparatus has a filtering unit 54 for filtering theused salt 16 from the molten salt electrolysis unit 20 to filter outnuclear fuel materials 19 from the used salt 16 to provide the filteredsalt 18. The filtering unit 54 may be a filtering device generally usedin chemical engineering and capable of separating the nuclear fuelmaterials 19 and the filtered salt 18 by subjecting the used salt 16 tofiltration.

The waste treatment apparatus has a filtered salt return line 55 forreturning the filtered salt 18 to the molten salt electrolysis unit 20.The filtered salt return line 55 may be of either a transfer pipe typeor a conveyor type.

Referring to FIG. 2, the molten salt electrolysis unit 20 has anelectrolytic vessel 20 a made of a low-carbon steel, an anode basket 21which is a mesh structure of a low-carbon steel or a stainless steel,placed in the electrolytic vessel 20 a, and a driving device 96. Thebasket 21 is driven for rotation in a molten salt 24 contained in theelectrolytic vessel 20 a by the driving device 96 to promoteelectrolytic reaction by moving the radioactive contaminated waste 10contained in the basket 21 relative to the molten salt 24.

The anode basket 21 containing the radioactive contaminated waste 10contaminated with nuclear fuel materials is immersed in the molten salt24. A cathode 23 of a low-carbon steel is immersed in the molten salt24. A dc power supply 25 has a positive electrode and a negativeelectrode connected to the anode basket 21 and the cathode 23,respectively. In FIG. 2, indicated at 26 is a cathodic deposit and at 27is sludge.

The molten salt 24 is an electrolyte prepared by melting one of chemicalcompounds including an alkali metal chloride, an alkaline earth metalchloride, an alkali metal fluoride, an alkaline earth metal fluoride, achloride or fluoride of an element included in the component elements ofthe waste 10, or a mixture of some of those chemical compounds, andkeeping the molten salt at a temperature not lower than its meltingpoint.

Referring to FIG. 3, the evaporating unit 59 has a melting crucible 70for heating the decontaminated waste 12 decontaminated by the moltensalt electrolysis unit 20 and soiled with the salt 24, and an inductionheating coil 71 surrounding the melting crucible 70. The decontaminatedwaste 12 contained in the melting crucible 70 is heated. Consequently,the decontaminated waste 12 melts into molten waste 72 and the salt 24adhering to the decontaminated waste 12 evaporates in a gas phase. Thesalt 24 in a gas phase flows in the direction of the arrows 73 and isrecovered to obtain the recovered salt 15 in a liquid phase.

A method of treating the radioactive contaminated waste 10 from anuclear fuel handling facility to be carried out by the waste treatmentapparatus in the first embodiment shown in FIGS. 1 to 3 will bedescribed with reference to FIGS. 1 to 4.

Referring to FIG. 4, a molten salt electrolysis process 11 puts theradioactive contaminated waste 10 from the nuclear fuel handlingfacility in the anode basket 21 and immerses the anode basket 21 in themolten salt 24 contained in the electrolytic vessel 20 a of the moltensalt electrolysis unit 20. A current is supplied through the radioactivecontaminated waste 10 functioning as an anode, and the cathode 23 todissolve electrochemically a surface layer of the radioactivecontaminated waste 10 contaminated with nuclear fuel materials in themolten salt 24 to provide decontaminated waste 12. When a dc voltage isapplied across the anode basket 21 and the cathode 23 by the dc powersupply 25, the radioactive contaminated waste 10 functions as an anode,and the surface layer of the radioactive contaminated waste 10 dissolvesin the molten salt 24. Consequently, the nuclear fuel materials adheringto the surface of the contaminated waste 10 fall into the molten salt24, and sludge of the nuclear fuel materials deposits on the bottom ofthe electrolytic vessel 20 a of the molten salt electrolysis unit 20.Ions of the component metals of the radioactive contaminated waste 10are reduced and cathodic deposit 26 deposits on the cathode 23.

The decontaminated waste 12 is soiled with the salt 24 used by themolten salt electrolysis process 11. The salt 24 adhering to thedecontaminated waste 12 is removed from the decontaminated waste 12 byan evaporation process 13 using the evaporating unit 59. The evaporationprocess 13 heats the decontaminated waste 12 at a temperature not lowerthan the melting point of the salt 24 in an environment of theatmospheric pressure or a reduced pressure to evaporate the salt 24 fromthe decontaminated waste 12. Thus clean waste 14 is obtained. Therecovered salt 15 is returned through the recovered salt return line 53to the molten salt electrolysis unit 20 and is reused for the moltensalt electrolysis process 11. Thus, the salt 15 is removed from thedecontaminated waste 12 to obtain the clean waste 14. In the evaporationprocess 13, the decontaminated waste 12 can be melted to reduce the sameto a metal ingot by heating the decontaminated waste 12 at a temperaturehigher than its melting point during or after the removal of the salt 24from the decontaminated waste 12.

The used salt 16 used in the molten salt electrolysis process 11contains sludge of the nuclear fuel materials 19 removed from theradioactive contaminated waste 10. A filtering process 17 filters outthe sludge from the used salt 16 by the filtering unit 54. The filteredsalt 18 thus filtered by the filtering unit 54 is returned through thefiltered salt return line 55 to the molten salt electrolysis unit 20 andis reused for the molten salt electrolysis process 11.

A molten salt electrolysis unit 20 in a first modification of the moltensalt electrolysis unit 20 shown in FIG. 2 will be described withreference to FIG. 5, in which parts like or corresponding to those ofthe molten salt electrolysis unit 20 shown in FIG. 2 are designated bythe same reference characters and the description thereof will beomitted. The molten salt electrolysis unit 20 shown in FIG. 5 isprovided with a liquid metal 28 instead of the solid cathode 23 shown inFIG. 2. The liquid metal 28 serves as a cathode. The liquid metal 28 isin a liquid phase at the temperature of the melting point of the moltensalt 24. The liquid metal 28 is contained in an electrically insulatingceramic pot 29, and the ceramic pot 29 containing the liquid metal 28 isimmersed in the molten salt 24. A cathode wire 30 has one end dipped inthe liquid metal 28 and the other end connected to the dc power supply25. The liquid metal 28 may be stirred by a stirring device to promotethe mixing of the cathodic deposit deposited on the surface of theliquid metal 28 with the liquid metal 28. The cathode wire 30 isextended through an electrically insulating ceramic tube 31 to insulatethe same from the molten salt 24. Ions of the component metals of theradioactive contaminated waste 10 are reduced on the surface of theliquid metal 28 and the cathode deposit is deposited on the surface ofthe liquid metal 28.

A molten salt electrolysis unit 20 in a second modification of themolten salt electrolysis unit 20 shown in FIG. 2 will be described withreference to FIG. 6, in which parts like or corresponding to those ofthe molten salt electrolysis unit 20 shown, in FIG. 2 are designated bythe same reference characters and the description thereof will beomitted. The molten salt electrolysis unit 20 shown in FIG. 6 isprovided with actuators 36 and 37 for vibrating the anode basket 21. Theanode basket 21 is held by an anode basket holding bar 38. The actuator36 vibrates the anode basket holding bar 38 in vertical directions, andthe actuator 47 vibrates the same in horizontal directions. Theactuators 36 and 37 are used selectively to vibrate the anode basketholding bar 38 at an optional frequency in horizontal directions,vertical directions or in both vertical and horizontal directions topromote the separation of the nuclear fuel material from the surface ofthe radioactive contaminated waste 10.

A molten salt electrolysis unit 20 in a third modification of the moltensalt electrolysis unit 20 shown in FIG. 2 will be described withreference to FIG. 7, in which parts like or corresponding to those ofthe molten salt electrolysis unit 20 shown in FIG. 2 are designated bythe same reference characters and the description thereof will beomitted. The molten salt electrolysis unit 20 shown in FIG. 7 isprovided with a cleaning device for cleaning the surface of theradioactive contaminated waste 10 in the molten salt 24. The cleaningdevice has a molten salt suction pipe 40, a molten salt jetting pipe 41provided with a molten salt jetting nozzle 42, and a pump 39. The moltensalt suction pipe 40 and the molten salt jetting pipe 41 are connectedto the inlet port and the outlet port of the pump 39, respectively. Thepump 39 operates to suck the molten salt 24 through the molten saltsuction pipe 40 and to clean the radioactive contaminated waste 10contained in the anode basket 21 by jetting the molten salt 24 throughthe molten salt jetting nozzle 42 against the radioactive contaminatedwaste 10. In FIG. 7 the arrows 43 indicate the flow of the molten salt24.

As apparent from the foregoing description, the waste treatmentapparatus in the first embodiment decontaminates the radioactivecontaminated waste 10 contaminated with the nuclear fuel materials bythe molten salt electrolysis unit 20, removes the salt 24 adhering tothe decontaminated waste 12 by heating the decontaminated waste 12 inthe environment of the atmospheric pressure or a reduced pressure toevaporate the salt 24 by the evaporating unit 59. Thus, the salt 24adhering to the decontaminated waste 12 can easily be removed from thedecontaminated waste 12 to obtain the clean waste 14. The recovered salt15 recovered by the evaporating unit 59 can be returned through therecovered salt return line 53 to the molten salt electrolysis unit 20 toreuse the same. The used salt 16 is filtered and the filtered salt 18can be returned through the filtered salt return line 55 to the moltensalt electrolysis unit 22 to reuse the same.

A waste treatment apparatus in a second embodiment according to thepresent invention for treating radioactive contaminated waste from anuclear fuel handling facility will be described hereinafter. The wastetreatment apparatus in the second embodiment is a modification of thewaste treatment apparatus in the first embodiment. Parts of the wastetreatment apparatus in the second embodiment like or corresponding tothose of the waste treatment apparatus in the first embodiment aredesignated by the same reference characters and the description thereofwill be omitted.

Referring to FIG. 8, the waste treatment apparatus in the secondembodiment is provided with a cleaning unit 56 instead of theevaporating unit 59 of the waste treatment apparatus in the firstembodiment, and is provided additionally with an evaporative drying unit57 and a cleaning liquid return line 58. The cleaning unit 56 cleans thedecontaminated waste 12 with a cleaning liquid, such as water. Therecovered cleaning liquid recovered by the evaporative drying unit 57 isreturned through the cleaning liquid return line 58 to the cleaning unit56. The recovered salt 15 recovered by the evaporative drying unit 57 isreturned through the recovered salt return line 53 to the molten saltelectrolysis unit 20 to reuse the same.

Referring to FIG. 9, the cleaning unit 56 has a filter 74 for filteringthe cleaning liquid, and a pump 75 for spraying the filtered cleaningliquid on the decontaminated waste 12 decontaminated by the molten saltelectrolysis unit 20.

A waste treatment method using the waste treatment apparatus shown inFIGS. 8 and 9 will be described with reference to FIGS. 8 to 10. Asshown in FIG. 10, the waste treatment method has a cleaning process 32instead of the waste treatment method shown in FIG. 4. The cleaningprocess 32 cleans the decontaminated waste 12 decontaminated by themolten salt electrolysis process 11 of the salt 24 adhering to thedecontaminated waste 12 with a cleaning liquid containing at least oneof liquids including water, a nitric acid solution, a sulfuric acidsolution and a hydrochloric acid solution. The used cleaning liquid 33containing the salt 24 and discharged from the cleaning unit 56 issubjected to evaporation by the evaporative drying unit 57 to recoverthe salt 24 by evaporative drying. The recovered salt 15 is returned tothe molten salt electrolysis unit 20 to reuse the same in the moltensalt electrolysis process 11. The cleaning liquid 35 recovered by theevaporative drying process 34 is returned through the cleaning liquidreturn line 58 to the cleaning unit 56 to reuse the same in the cleaningprocess 32.

As apparent from the foregoing description, the waste treatmentapparatus in the second embodiment is capable of readily removing thesalt 24 adhering to the decontaminated waste 12 by the cleaning unit 56after the radioactive contaminated waste 10 contaminated with thenuclear fuel materials has been decontaminated by the molten saltelectrolysis unit 20. The recovered salt 15 recovered by the evaporativedrying unit 57 is returned through the recovered salt return line 53 tothe molten salt electrolysis unit 20 and can be reused. The used salt 16used by the molten salt electrolysis unit 20 is filtered by thefiltering unit 54 to recycle the filtered salt 18. The filtered salt 18is returned through the filtered salt return line 55 to the molten saltelectrolysis unit 20 and can be reused in the molten salt electrolysisprocess 11. The cleaning liquid 35 recovered by the evaporative dryingunit 57 is returned through the cleaning liquid return line 58 to thecleaning unit 56. Thus, the cleaning liquid 35 can efficiently be reusedand hence additional effluents are not produced.

A waste treatment apparatus in a third embodiment according to thepresent invention for treating radioactive contaminated waste from anuclear fuel handling facility, and a waste treatment method to becarried out by the same waste treatment apparatus will be describedhereinafter. The waste treatment apparatus in the third embodiment is amodification of the waste treatment apparatus in the second embodiment.Parts of the waste treatment apparatus in the third embodiment like orcorresponding to those of the waste treatment apparatus in the secondembodiment are designated by the same reference characters and thedescription thereof will be omitted.

Referring to FIG. 11, the waste treatment apparatus in the thirdembodiment has a reducing unit 60 disposed on the upstream side of themolten salt electrolysis unit 20. When the nuclear fuel materialsadhering to the waste 10 are uranium ore or oxides, the reducing unit 60reduces the nuclear fuel materials prior to the treatment of the waste10 by the molten salt electrolysis unit 20 for the efficient molten saltelectrolysis of the waste 10.

Referring to FIG. 12, the reducing unit 60 has a reaction vessel 45containing a molten salt 47, a meshed waste container 46 placed in thereaction vessel 45 to contain the waste 10, and a stirring device 48inserted in the waste container 46. A reducing agent 49 is supplied intothe reaction vessel 45. The reducing agent 49 is lithium (Li), magnesium(Mg) or calcium (Ca). Preferably, the reducing agent 49 is Li. The waste10 from a nuclear fuel material handling facility is put into the wastecontainer 46. The reducing agent 49, such as Li, comes into directcontact with the waste 10 for reducing reaction.

A waste treatment method to be carried out by the waste treatmentapparatus in the third embodiment shown in FIGS. 11 and 12 will bedescribed with reference to FIGS. 11 to 13. The waste treatment methodcomprises a reducing process 44 in addition to the processes of thewaste treatment method to be carried out by the waste treatmentapparatus in the second embodiment. The waste treatment method carriesout the reducing process 44 by the reducing unit 60 before the moltensalt electrolysis process 11. The reducing process 44 reduces thenuclear fuel materials adhering to the waste 10 to metals through thedirect interaction of the reducing agent 49 and the nuclear fuelmaterials. The waste 10 thus treated by the reducing process 44 issubjected to processes entirely the same as those of the waste treatmentmethod shown in FIG. 10.

FIG. 14 shows a reducing unit 60 in a first modification of the reducingunit 60 shown in FIG. 12. The reducing unit 60 in the first modificationcomprises, in addition to the components of the reducing unit 60 shownin FIG. 12, a reducing agent regenerating device for regenerating thereducing agent. The reducing agent regenerating device comprises acathode 50 inserted in the waste container 46, an anode 52 (carbonelectrode) inserted in the reaction vessel 45, and a power supply 51 forapplying a voltage across the cathode 50 and the anode 52. Suppose thatthe reducing agent is Li. Lithium oxide (Li₂O) is produced by thereduction reaction of Li and the nuclear fuel materials adhering to thewaste 10, and Li₂O disperses in the waste container 46. Part of the Li₂Ois converted into Li and O at the cathode 50. Part of the thusregenerated Li is used for reduction and the rest disperses in the wastecontainer 46. The Li dispersed in the waste container 46 does notcontribute to reduction and hence efficient reduction cannot beachieved. A voltage that will not decompose a molten salt 46 containedin the reaction vessel 45, for example about 3 V, is applied across thecathode 50 and the anode 52 by the power supply 51. Consequently, Li,i.e., the reducing agent 49, supplied into the reaction vessel 45penetrates the waste container 46 gradually, comes into contact with thenuclear fuel materials adhering to the waste 10 and reducing reactionprogresses. Oxygen (O) generated when Li is regenerated at the cathode50 disperses outside the waste container 46. The stirring device 48disposed in the waste container 46 promotes the dispersion of O and thesupply of O to the anode 52. The following electrode reactions occur atthe electrodes during the foregoing processes.

Li₂O→2Li⁺+O²⁻

Anode: 2O²⁻+C→CO₂+4e⁻

 O²⁻+C→CO+2e⁻

Cathode: Li⁺+e⁻→Li

After the completion of the reducing reaction, the waste container 46 israised and pulled out of the molten salt 47 contained in the reactionvessel 45.

FIG. 15 shows a reducing unit 60 in a second modification of thereducing unit 60 shown in FIG. 12. The reducing unit 60 shown in FIG. 15has a cathode 61 and an anode 62 immersed in a molten salt 47. A voltageis applied across the cathode 61 and the anode 62 by a power supply 73to reduce oxides dispersed in a molten salt 47 to metals by electrolyticreduction. A reducing reaction progresses in the molten salt 47contained in a reaction vessel 45 of the reducing unit 60. Oxides, i.e.,nuclear fuel materials, are reduced to metals U and TRU at the cathode61, and O generated at the cathode 61 disperses outside a wastecontainer 46. A stirring device 48 disposed in the waste container 46promotes the dispersion of O and promotes the supply of O to the anode62 (carbon electrode). The following reactions occur at the electrodesduring the foregoing processes.

UO₂+4e→U+2O²⁻

Anode: 2O²⁻+C→CO₂+4e⁻

 O²⁻+C→CO+2e⁻

Cathode: U⁴⁺+4e⁻→U

After the completion of the reducing reaction, the waste container 46 israised and pulled out of the molten salt 47 contained in the reactionvessel 45.

A waste treatment apparatus in a fourth embodiment according to thepresent invention for treating radioactive contaminated waste from anuclear fuel handling facility, and a waste treatment method to becarried out by the same waste treatment apparatus will be describedhereinafter. The waste treatment apparatus in the fourth embodiment issuitable for treating radioactive contaminated waste when theradioactive contaminated waste is an adsorbent, such as NaF, and thenuclear fuel materials adhering to the adsorbent are fluorides, such asUF₆, UF₄ and UO₂F₂.

Referring to FIG. 16, the waste treatment apparatus comprises a reducingunit 60 for reducing waste 100, a thermal melting unit 64 connected tothe reducing unit 60, a molten salt electrolysis unit 65 connected tothe thermal melting unit 64, a cleaning unit 66 connected to the moltensalt electrolysis unit 65, a evaporative drying unit 67 connected to thecleaning unit 66 and an oxidizing unit 68 connected to the cleaning unit66.

The reducing unit 60 reduces radioactive contaminated waste 100. Thethermal melting unit 64 heats and melts the reduced waste 101 providedby reducing the radioactive contaminated waste 100 by the reducing unit60. The molten salt electrolysis unit 65 subjects a molten salt 102,i.e., the molten waste provided by the thermal melting unit 64 toelectrolysis. Thus, the molten waste prepared by melting the reducedwaste 101 produced by reducing the radioactive contaminated waste 100 bythe reducing unit 60 is used as the molten salt 102 for electrolysis.The cleaning unit 66 separates nuclear fuel materials (uranium metal)and an adsorbent (NaF) contained in a cathodic deposit 76 deposited onthe cathode of the molten salt electrolysis unit 65. The evaporativedrying unit 67 processes a used cleaning liquid 77 used by the cleaningunit 66 for evaporative drying to recover the adsorbent (NaF) dissolvedin the used cleaning liquid 77. A cleaning liquid 78 recovered byevaporation is returned through a recovered cleaning liquid return line79 to the cleaning unit 66 and is reused. The nuclear fuel materials(uranium metal) 80 separated from the adsorbent by the cleaning unit 66is oxidized by the oxidizing unit 68, and oxides (Uranium oxide) 81 thusproduced by the oxidizing unit 68 are collected.

As shown in FIG. 17, the molten salt electrolysis unit 65 comprises areaction vessel 85 for containing the molten salt 102 prepared bymelting the reduced waste 101, an anode 82 and a cathode disposed in thereaction vessel 85, and a power supply 84 for applying a voltage acrossthe anode 82 and the cathode 83.

A waste treatment method to be carried out by the waste treatmentapparatus in the fourth embodiment shown in FIGS. 16 and 17 will bedescribed with reference to FIGS. 16 to 18. A reducing process 86processes the radioactive contaminated waste 100, i.e., the adsorbent(NaF) contaminated with the nuclear fuel materials, such as UF₆, UF₄ andUO₂F₂, to reduce, for example, UF₆ (uranium hexafluoride) to UF₄(uranium tetrafluoride). More concretely, a reducing gas, such ashydrogen gas, argon gas or phosgene gas, is spouted against adsorbentparticles to reduce the nuclear fuel materials.

A thermal melting and salt-processing process 87 heats and melts thereduced waste 101, i.e., the adsorbent containing the reduced nuclearfuel materials by the thermal melting unit 64 and adds a fluoride or ahydride having the same cations as those of the reduced waste 101 to themolten waste 101 to produce a molten salt 102 having a low meltingpoint. The chloride having the same cations as those of the reducedwaste 101 is, for example, NaCl. When NaF and NaCl are mixed, a eutecticof NaF—NaCl having a melting point of 600° C. is produced. The meltingpoint of this eutectic is lower than the melting point of 992° C. of NaFby 390° C.

In a molten salt electrolysis process 88, the anode 82 and the cathode83 of the molten salt electrolysis unit 65 are immersed in the moltensalt 102, and a voltage is applied across the anode 82 and the cathode83 to reduce UF₄ to uranium metal. A cathodic deposit 76 containinguranium metal, NaF and NaCl is deposited on the cathode 83. The cathodicdeposit 76 is recovered from the cathode 83. A cleaning process 89cleans the cathodic deposit 76 with a cleaning liquid, such as water toseparate uranium metal from other components of the cathodic deposit 76.An oxidizing process 90 oxidizes the thus recovered uranium metal touranium oxide by the oxidizing unit 68. The uranium oxide is stable inthe atmosphere. An evaporative drying process 91 heats and evaporatesthe used cleaning liquid 77 containing NaF by the evaporative dryingunit 67 to recover the NaF. An evaporated cleaning liquid 78 is returnedto the cleaning unit 66 and is reused.

As shown in FIG. 19, a waste treatment apparatus in a modification ofthe waste treatment apparatus shown in FIG. 16 heats and melts theradioactive contaminated waste 100 before the reducing process 86 toproduce a molten salt 92, blows a reducing gas 93, such as hydrogen gas,argon gas or phosgene gas, through a nozzle 95 into the molten salt 92contained in a vessel 94 to reduce the nuclear fuel materials within themolten salt 92. The nuclear fuel materials may be reduced byelectrolytic reduction by immersing an anode and a cathode in the moltensalt 92 and applying a voltage across the anode and the cathode. Achloride or the like is added to the molten salt after reduction and thereduced molten salt 92 is subjected to electrolysis by the molten saltelectrolysis unit 65.

As apparent from the foregoing description, according to the presentinvention, electrically conductive waste contaminated with nuclear fuelmaterials from a nuclear fuel handling facility is immersed in a moltensalt, the waste is connected to an anode and a surface layer of thewaste is dissolved electrochemically in the molten salt. Thus, thenuclear fuel materials adhering to the waste can easily be removed.

Since the electrical resistance of the molten salt is very low ascompared with that of an electrolytic water solution, an electriccurrent flows uniformly over the surface of the waste. Consequently, thewaste having a complicated shape, which is difficult to decontaminate byconventional techniques, can surely be decontaminated. Since theelectrical resistance of the molten salt is low, a large current can besupplied through the molten salt without entailing abnormal heatgeneration to increase the process speed. The molten salt electrolysisprocess is safe because hydrogen is not generated at the cathode whenthe molten salt is used for the electrolysis.

The sludge of the nuclear fuel material accumulated in the molten saltcan satisfactorily be separated from the molten salt by filtrationbecause the surface tension of the molten salt is lower than that of anaqueous solution. The nuclear fuel material dissolved in the molten saltcan be recovered in a cathodic deposit. The molten salt can be reusedeven if some nuclear fuel material dissolved in the molten salt remainsin the molten salt.

The waste, such as the absorbent used in the nuclear fuel materialhandling facility, can be easily treated by reducing the nuclear fuelmaterials to metals, heating and melting the metals and the waste, andrecovering the metals on the cathode in the molten salt electrolysis.

Although the invention has been described in its preferred form with acertain degree of particularity, obviously many changes and variationsare possible therein. It is therefore to be understood that the presentinvention may be practiced otherwise than as specifically describedherein without departing from the scope and spirit thereof.

What is claimed is:
 1. A method of treating electrically conductivewaste contaminated with nuclear fuel materials from a nuclear fuelhandling facility, which comprises: a molten salt electrolysis processfor removing the nuclear fuel materials adhering to a surface of thewaste by immersing the waste in a molten salt to dissolve a surfacelayer of the waste electrochemically in molten salt without dissolvingthe nuclear fuel materials so as to provide a decontaminatedelectrically conductive waste; and a filtering process for filtering themolten salt used in the molten salt electrolysis process to extract thenuclear fuel materials removed from the surface of the waste andaccumulated in the molten salt from the molten salt; wherein the moltensalt filtered in the filtering process is reused in the molten saltelectrolysis process.
 2. The method according to claim 1 furthercomprising an evaporation process for removing the molten salt adheringto a surface of the waste processed by the molten salt electrolysisprocess and taken out of the molten salt by heating the waste so thatthe molten salt adhering thereto evaporates; wherein the molten saltrecovered in the evaporation process is reused in the molten saltelectrolysis process.
 3. The method according to claim 1 furthercomprising: a cleaning process for removing the molten salt adhering tothe waste processed by the molten salt electrolysis process and takenout of the molten salt by a cleaning liquid; and an evaporative dryingprocess for drying the molten salt contained in the cleaning liquid byevaporating the cleaning liquid used in the cleaning process; whereinthe molten salt recovered in the evaporative drying process is reused inthe molten salt electrolysis process, and the cleaning liquid evaporatedin the evaporative drying process is reused in the cleaning process. 4.The method according to claim 1, wherein the molten salt and the wasteimmersed in the molten salt are moved relative to each other in themolten salt electrolysis process to remove the nuclear fuel materialsfrom the surface of the waste.
 5. The method according to claim 4,wherein, in the molten salt electrolysis process, the waste is containedin a basket serving as an electrode for an electrolysis and the basketis vibrated in the molten salt.
 6. The method according to claim 4,wherein, in the molten salt electrolysis process, the waste is containedin a basket serving as an electrode for an electrolysis and the basketis rotated in the molten salt.
 7. The method according to claim 4,wherein, in the molten salt electrolysis process, the molten salt isspouted against the waste immersed in the molten salt.
 8. The methodaccording to claim 1, wherein, a liquid metal, which is in a liquidphase at a temperature high enough to maintain the molten salt in amolten state, is placed in the molten salt as an electrode for themolten salt electrolysis process.
 9. The method according to claim 1further comprising a reducing process for reducing the nuclear fuelmaterials to metals before subjecting the waste to the molten saltelectrolysis process when the nuclear fuel materials are oxides.
 10. Themethod according to claim 9, wherein, in the reducing process, thenuclear fuel materials are reduced to metals by making the nuclear fuelmaterials react with a reducing agent.
 11. The method according to claim10, wherein the reducing process comprises: immersing the wastecontaminated with the nuclear fuel materials in a reducing molten salt;supplying a reducing agent into the reducing molten salt; and applying avoltage that will not cause a decomposition of the reducing molten saltacross an anode and a cathode immersed in the reducing molten salt toregenerate the reducing agent reacted with the nuclear fuel materials.12. The method according to claim 9, wherein the reducing processcomprises: immersing the waste contaminated with the nuclear fuel oxidesin a reducing molten salt; and reducing the nuclear fuel materials tometals by applying a voltage across an anode and a cathode immersed inthe reducing molten salt for an electrolytic reduction.